Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from, May 14 through May 27, 2004. The last biweekly notice was published on May 25, 2004 (69 FR 29761).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60-day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30-day comment period should circumstances change during the 30-day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to pdr@nrc.gov.

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of amendment request: April 26, 2004.

Description of amendment request: The proposed amendment would revise the Completion Time for Required Action A.1 of Technical Specification 3.8.7, “Inverters—Operating,” from the current 24 hours for a Division 1 or 2 Nuclear System Protection System (NSPS) inverter inoperable to 7 days.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed TS change revises the Completion Time for Required Action A.1 associated with the Division 1 and 2 NSPS inverters. Specifically, the proposed action allows continued unit operation, for up to 7 days, with an inoperable Division 1 or 2 NSPS inverter.

The proposed change does not affect the design of the NSPS inverters, the operational characteristics or function of the inverters, the interfaces between the inverters and other plant systems, or the reliability of the inverters. An inoperable NSPS inverter is not considered as an initiator of any analyzed event. In addition, Required Actions and the associated Completion Times are not initiators of any previously evaluated accidents. Extending the Completion Time for an inoperable NSPS inverter would not have a significant impact on the frequency of occurrence for any accident previously evaluated. The proposed change will not result in changes to the plant activities associated with NSPS inverter maintenance, but rather will allow increased flexibility in the scheduling and performance of preventive maintenance. Therefore, this change will not significantly increase the probability of occurrence of any event previously analyzed.

The consequences of a previously analyzed event are dependent on the initial conditions assumed in the analysis, the availability and successful functioning of equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. With an NSPS inverter inoperable, the affected instrument bus is capable of being fed from its dedicated safety-related alternate power supply, which is powered from a Class 1E 480 VAC bus through a step-down transformer and an isolation transformer. In the event of a Loss of Offsite Power (LOOP), the affected instrument bus will experience a momentary loss of power until the associated diesel generator (DG) re-energizes the 480 VAC bus. A LOOP with an inoperable NSPS inverter (i.e., instrument bus being powered by its alternate power supply) will result in a loss of power to the associated instrument bus until the associate DG re-energizes the Class 1E 480 VAC bus. All instruments supplied by the instrument bus would be restored with no adverse impact to the unit because no other instrument channels in the opposite train would be expected to be inoperable or in a tripped condition during this time, with the exception of routine surveillances. In the event of a failure to re-energize the 480 VAC bus or of a transformer failure, the most significant impact on the unit is the failure of one train of Engineered Safety Feature (ESF) equipment to actuate. In this condition, the redundant train of ESF equipment will automatically actuate to mitigate the accident, and the affected unit would remain within the bounds of the accident analyses. In addition, there would be no adverse impact to the unit because no other instrument channels in the opposite train would be expected to be inoperable or in a tripped condition during this time, with the exception of routine surveillances.

To fully evaluate the effect of the proposed NSPS inverter Completion Time extension, probabilistic risk assessment (PRA) methods and a deterministic analysis were utilized. The Incremental Conditional Core Damage Probability (ICCDP) and Incremental Conditional Large Early Release Probability (ICLERP) for each inverter division are sufficiently below the regulatory guidelines to be able to call the risk change small. Hence, the guidelines of Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed Decision-Making: Technical Specifications,” for the increased inverter Completion Time have been met. Furthermore, the evaluation of changes in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) due to the expected increased inverter unavailability, as mitigated by the compensating measures assumed in the analysis, have been shown to meet the risk significance criteria of Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” with substantial margin. This calculation supports the increase in the Division 1 and 2 inverter Completion Times from a quantitative risk-informed perspective consistent with the plant operational and maintenance practices. Therefore, the request for extending the Completion Time will not significantly increase the consequences of an accident previously evaluated.

In summary, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed action does not involve physical alteration of the station. No new equipment is being introduced, and installed equipment is not being operated in a new or different manner. There is no change being made to the parameters within which CPS is operated. There are no setpoints at which protective or mitigative actions are initiated that are affected by this proposed action. The use of the alternative Class 1E power source for the instrument bus is consistent with the CPS plant design. The change does not alter assumptions made in the safety analysis. This proposed action will not alter the manner in which equipment operation is initiated, nor will the function demands on credited equipment be changed. No alteration in the procedures, which ensure the unit remains within analyzed limits, is proposed, and no change is being made to procedures relied upon to respond to an off-normal event. As such, no new failure modes are being introduced.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a marge of safety?

Response: No.

Margins of safety are established in the design of components, the configuration of components to meet certain performance parameters, and in the establishment of setpoints to initiate alarms or actions. There is no change in the design of the affected systems, no alteration of the setpoints at which alarms or actions are initiated, and no change in plant configuration from original design. With one of the required instrument buses being powered from the alternate class 1E power supply, there is no significant reduction in the margin of safety. Testing of the DGs and associated electrical distribution equipment provides confidence that the DGs will start and provide power to the associated equipment in the unlikely event of a LOOP during the extended 7-day Completion Time.

Applicable regulatory requirements will continue to be met, adequate defense-in-depth will be maintained, sufficient safety margins will be maintained, and any increases in risk are small and consistent with the NRC Safety Goal Policy Statement (Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986, as interpreted by NRC Regulatory Guides 1.174 and1.177). Furthermore, increases in risk posed by potential combinations of equipment out of service during the proposed NSPS inverter extended Completion Time will be managed under a configuration risk management program (CRMP) consistent with 10CFR50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants.”, paragraph (a)(4).

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60666.

NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: April 30, 2004.

Description of amendment request: The proposed amendments would incorporate the oscillation power range monitor (OPRM) instrumentation into the technical specifications (TS). The proposed changes would revise: (1) TS 3.3.1.3, “Oscillation Power Range Monitor (OPRM) Instrumentation,” to insert a new TS section for the OPRM instrumentation, (2) TS 3.4.1, “Recirculation Loops Operating,” to delete the current thermal hydraulic instability administrative requirements, and (3) TS 5.6.5, “Core Operating Limits Report (COLR),” to add the appropriate references for the OPRM trip setpoints and methodology.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. This proposed change has no impact on any of the existing neutron monitoring functions.

Activation of the OPRM scram function will replace the current methods that require operators to insert an immediate manual reactor scram in certain reactor operating regions where thermal hydraulic instabilities could potentially occur. While these regions will continue to be avoided during normal operation, certain transients, such as a reduction in reactor recirculation flow, could place the reactor in these regions. During these transient conditions, with the OPRM instrumentation scram function activated, an immediate manual scram will no longer be required. This may potentially cause a marginal increase in the probability of occurrence of an instability event. This potential increase in probability is acceptable because the OPRM function will automatically detect the instability condition and initiate a reactor scram before the Minimum Critical Power Ratio (MCPR) Safety Limit is reached. Consequences of the potential instability event are reduced because of the more reliable automatic detection and suppression of an instability event, and the elimination of dependence on the manual operation actions. Operators will continue to monitor for indications of thermal hydraulic instability when the reactor is operating in regions of potential instability as a backup to the OPRM instrumentation. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The proposed changes replace procedural actions that were established to avoid operating conditions where reactor instabilities might occur with an NRC approved automatic detect and suppress function (i.e., OPRM).

Potential failure in the OPRM trip function could result in either a failure to take the required mitigating action or an unintended reactor scram. These are the same potential effects of failure of the operator to take the correct appropriate action under the current procedural actions. The effects of failures of the OPRM equipment are limited to reduced or failed mitigation, but such failure cannot cause an instability event or other type of accident.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The OPRM trip function is being implemented to automate the detection and subsequent suppression of an instability event prior to exceeding the MCPR Safety Limit. The OPRM trip provides a trip output of the same type as currently used for the [average power range monitor] APRM. Its failure modes and types are identical to those for the present APRM output. Since the MCPR Safety Limit will not be exceeded as a result of an instability event following implementation of the OPRM trip function, it is concluded that the proposed change does not reduce the margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: April 13, 2004.

Description of amendment request: The proposed amendment deletes requirements from the Technical Specifications to maintain hydrogen recombiners and hydrogen and oxygen monitors. Licensees were generally required to implement upgrades as described in NUREG-0737, “Clarification of TMI [Three Mile Island] Action Plan Requirements,” and Regulatory Guide (RG) 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.” Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the technical specifications (TS) for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, “Standards for combustible gas control system in light-water-cooled power reactors,” eliminated the requirements for hydrogen recombiners and relaxed safety classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The NRC staff issued a notice of availability of a model no significant hazards consideration determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated March 4, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The revised 10 CFR 50.44 no longer defines a design-basis loss-of-coolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant accident sequences that could threaten containment integrity.

With the elimination of the design-basis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate design-basis accidents and, therefore, the hydrogen monitors do not meet the definition of a safety-related component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for design-basis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44 the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond design-basis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment.

The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3, classification of the oxygen monitors as Category 2 and removal of the hydrogen and oxygen monitors from TS will not prevent an accident management strategy through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or the consequences of any accident previously evaluated.

Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a design-basis hydrogen release. The hydrogen recombiner and hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the pre-accident state of the reactor core or post accident confinement of radionuclides within the containment building.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that this hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related hydrogen monitors. Category 2 oxygen monitors are adequate to verify the status of an inerted containment.

Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safety-related oxygen monitors. Removal of hydrogen and oxygen monitoring from TS will not result in a significant reduction in their functionality, reliability, and availability.

Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: May 5, 2004.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TS) for instrumentation setpoints, allowable values, and calibration requirements based on updated calculations and reviews, and add a definition of “annual” frequency for use in the TS.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensees have provided their analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed new DVR [degraded voltage relay] voltage and minimum time delay Allowable Values are more restrictive than the existing TS limits. The proposed new DVR maximum time delay is based on the existing analytical limit, and is only increased to the extent permitted by the methods endorsed by Regulatory Guide (RG) 1.105. Annual channel calibrations are already performed, and adding them to TS ensures from a regulatory perspective that the relay drift is consistent with the setpoint calculations. The proposed new LVR [loss of voltage relay] voltage upper Allowable Value is based on a comprehensive EDG [emergency diesel generator] transient analysis, and is only increased to the extent permitted by the methods endorsed by Regulatory Guide (RG) 1.105. The proposed new LVR time delay allowable values are more restrictive than the existing TS limits, and are within the existing TS range of allowable values. Accident initial conditions, probability, and assumptions remain as previously analyzed. The remaining portions of the amendment request are administrative changes that will have no effect on operations of the relays. The Degraded Voltage and Loss of Voltage Relays are not accident initiators; therefore, a malfunction of these relays will have no significant effect on accident initiation frequency. The proposed changes do not invalidate the assumptions used in evaluating the radiological consequences of any accident. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed new DVR voltage and minimum time delay Allowable Values are more restrictive than the existing TS limits. The proposed new DVR maximum time delay is based on the existing analytical limit, and is only increased to the extent permitted by the methods endorsed by Regulatory Guide (RG) 1.105. Annual channel calibrations are already performed, and adding them to TS ensures from a regulatory perspective that the relay drift is consistent with the setpoint calculations. The proposed new LVR voltage upper Allowable Value is based on a comprehensive EDG transient analysis, and is only increased to the extent permitted by the methods endorsed by Regulatory Guide (RG) 1.105. The proposed new LVR time delay allowable values are more restrictive than the existing TS limits, and are within the existing TS range of allowable values. Accident initial conditions and assumptions remain as previously analyzed. The remaining portions of the amendment request are administrative changes that will have no effect on operations of the relays.

The proposed changes do not introduce any new or different accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

No. The proposed changes to the DVR Allowable Values will ensure an adequate margin of safety is maintained between the lowest allowable voltage setpoint and the highest per unit voltage required by safety-related equipment, while at the same time establishing an Allowable Value, not previously provided, that ensures a sufficient margin of safety between the highest allowable voltage setpoint and the lowest expected per unit source voltages.

The proposed changes to the DVR Allowable Values will ensure an adequate margin of safety is maintained between the longest allowable time delay and the longest time delay assumed by the accident analyses, while at the same time establishing an Allowable Value, not previously provided, that ensures a sufficient margin of safety between the shortest allowable time delay and the longest acceleration time for 4160 Volt continuously energized Safety Features Actuation System motors.

The proposed new LVR voltage upper Allowable Value is based on a comprehensive EDG transient analysis, and is only increased to the extent permitted by the methods endorsed by Regulatory Guide (RG) 1.105. In addition, the new Allowable Value reflects improvements in channel uncertainties that were made possible by upgrading the relays to solid state units.

The proposed new LVR time delay allowable values are more restrictive than the existing TS limits, and are within the existing TS range of allowable values.

A new requirement to perform an annual channel calibration of the Degraded Voltage and Loss of Voltage Relays is proposed. This new requirement to demonstrate proper channel operations will not adversely affect a margin of safety. The remaining changes are administrative, and will have no effect on margin of safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 21, 2004.

Description of amendment request: The proposed amendment would revise the following in the technical specifications (TSs): (1) adding a new figure (Figure 2-3) to the table of contents that shows the volume of Trisodium Phosphate (TSP) required over the operating cycle; (2) Section 2.3(4), “Emergency Core Cooling System—Trisodium Phosphate (TSP),” regarding volume and form of TSP; and (3) Section 3.6(2)d.(i), “Safety Injection and Containment Cooling Systems Tests,” regarding the surveillance requirement for TSP volume. The amendment also proposes modifications to the corresponding Basis of TS 2.3 and TS 3.6.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

There are no changes to the design or operation of the plant that could affect system, component, or accident functions as a result of deleting the requirement for the “dodecahydrate” form of TSP, or replacing the volume of active TSP required during Operating Modes 1 and 2 with an amount dependent upon HZP [hot zero power] CBC [critical boron concentration] as shown in Figure 2-3. All systems and components function as designed and the performance requirements have been evaluated and found to be acceptable. Hydrated TSP in the range of 45-57% moisture content will maintain pH ≥ 7.0 in the recirculation water following a LOCA [loss-of-coolant accident]. This function is maintained with the proposed change. Allowing the required volume of active TSP to decrease over the operating cycle as HZP CBC decreases will ensure that the pH of the containment sump is ≥ 7.0 yet provides additional margin for EEQ [equipment environmental qualification] concerns as containment sump pH is less likely to exceed 7.5.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of the proposed change. All systems, structures, and components previously required for mitigation of an event remain capable of fulfilling their intended design function with this change to the TS. The proposed change has no adverse effects on any safety-related systems or component and does not challenge the performance or integrity of any safety related system. The proposed change has evaluated the TSP configuration such that no new accident scenarios or single failures are introduced.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

Deleting the requirement for the “dodecahydrate” form of TSP and allowing the required volume of active TSP to decrease as HZP CBC decreases still ensures that the pH of the containment sump is ≥ 7.0. Hydrated TSP in the range of 45-57% moisture content will maintain pH ≥ 7.0 in the recirculation water following a LOCA. This change provides additional margin for EEQ concerns as containment sump pH is less likely to exceed 7.5. Therefore, this change does not involve a significant reduction in the margin of safety. Evaluations were made that indicate that the margin for pH control is not altered by the proposed changes. A TSP volume that is dependent on HZP CBC has been evaluated with respect to neutralization of all borated water and acid sources. These evaluations concluded that there would be no impact on pH control, and hence no reduction in the margin of safety related to post LOCA conditions.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 2004.

Description of amendment request: The proposed amendment would revise the reactor pressure vessel pressure-temperature limits and extend the validity of the limits to 32 effective full power years.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The revised curves are based on uprated fluence projections and are applicable for the service period up to 32 effective full power years (EFPY). There are no changes being made to the reactor coolant system (RCS) pressure boundary or to RCS material, design or construction standards. The proposed heatup and cooldown curves define limits that continue to ensure the prevention of nonductile failure of the RCS pressure boundary. The design-basis events that were evaluated have not changed. The modification of the heatup and cooldown curves does not alter any assumptions previously made in the radiological consequence evaluations since the integrity of the RCS pressure boundary is unaffected. Therefore, the proposed changes will not significantly increase the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Revisions to the heatup and cooldown curves do not involve any new components or plant procedures. The proposed changes do not create any new single failure or cause any systems, structures, or components to be operated beyond their design bases. Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed figures define the limits for ensuring prevention of nonductile failure for the reactor coolant system based on the methods described in 1989 ASME Code [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] Section XI Appendix G, 10 CFR 50 Appendix G, and ASME Code Cases N-640 and N-588. The effect of the change is to permit plant operation within different pressure-temperature limits, but still with adequate margin to assure the integrity of the reactor coolant system pressure boundary. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) The applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of application for amendments: December 15, 2003.

Brief Description of amendments: The amendments revise Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” for the condition of having one or more SDV vent or drain lines with one valve inoperable.

Date of issuance: May 17, 2004.

Effective date: May 17, 2004.

Amendment Nos.: 232 and 259.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12364).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 17, 2004.

No significant hazards consideration comments received: No.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of application for amendments: August 19, 2003, supplements dated October 23, 2003, and January 28, 2004.

Brief description of amendments: The amendments revised the Technical Specifications to modify the requirements for the containment pressure control system to eliminate a problem with circuit fluctuation as a result of electronic noise.

Date of issuance: May 12, 2004.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 214 and 208.

Renewed Facility Operating License Nos. NPF-35 and NPF-52: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 18, 2003 (68 FR 54749).

The supplements dated October 23, 2003, and January 28, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on September 18, 2003 (68 FR 54749).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 12, 2004.

No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: May 14, 2002, as supplemented by letters dated June 27, 2002, July 9, 2003, and April 7 and May 12, 2004.

Brief description of amendment: The amendment revises the Updated Safety Analysis Report (USAR) Appendix 3B and Sections 6.2.1.1.3.2.1, “Reactor Water Cleanup Break” and 6.2.1.2 “Containment Subcompartments” to change the method of analysis for high energy line breaks inside and outside of containment. The change will replace the current THREED code for room pressure-temperature analyses with the GOTHIC (Generation of Thermal-Hydraulic Information for Containments) code.

Date of issuance: May 20, 2004.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 139.

Facility Operating License No. NPF-47: The amendment revised the USAR Appendix 3B and Sections 6.2.1.1.3.2.1 and 6.2.1.2.2.

Date of initial notice in Federal Register: July 9, 2002 (67 FR 45563). The June 27, 2002, July 9, 2003, and April 7 and May 12, 2004, supplemental letters provided clarifying information that did not expand the scope of the original Federal Register notice or the staff's original proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 20, 2004.

No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 12, 2003, as supplemented by letter dated April 22, 2004.

Brief description of amendment: The amendment changes the heater acceptance criteria contained in surveillance requirements 4.6.6.1d.5, 4.7.6.1d.3, and 4.7.7d.4, performed to verify that the heat dissipated by the heaters is within a given band, for the shield building ventilation, control room ventilation, and controlled ventilation area systems, respectively. The changes increase the upper limit of the acceptance criteria from rated capacity plus 5 percent to rated capacity plus 10 percent and without any change for the lower limit of the band of rated capacity minus 10 percent.

Date of issuance: May 24, 2004.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 194.

Facility Operating License No. NPF-38: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: October 14, 2003 (68 FR 59217). The April 22, 2004, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 24, 2004.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of application for amendments: January 28, 2004, as supplemented May 3, 2004

Brief description of amendments: The amendments eliminated the requirements in BVPS-1 and BVPS-2 Technical Specifications (TSs) associated with hydrogen recombiners and relocate the requirements for hydrogen monitors to the Licensing Requirements Manuals.

Date of issuance: May 19, 2004.

Effective date: As of the date of issuance, and shall be implemented within 120 days.

Amendment Nos.: 259 and 142.

Facility Operating License Nos. DPR-66 and NPF-73: The amendments revised the TSs.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12370). The supplement dated May 3, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 19, 2004.

No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: July 14, 2003, as supplemented November 20, 2003, March 25, 2004, and April 27, 2004.

Brief description of amendment: The amendment allows a one-time increase in the completion time for restoring an inoperable nuclear services seawater system train to operable status.

Date of issuance: May 18, 2004.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment No.: 212.

Facility Operating License No. DPR-72: Amendment revises the License and Technical Specifications.

Date of initial notice in Federal Register: August 5, 2003 (68 FR 42644). The November 20, 2003, March 25, 2004, and April 27, 2004, supplements contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 18, 2004.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: September 26, 2003, as supplemented January 7, 2004.

Brief description of amendments: The amendments modify Technical Specifications (TS) requirements to adopt the provisions of Industry/TS Task Force (TSTF) change TSTF-359, “Increased Flexibility in Mode Restraints.”

Date of issuance: May 12, 2004.

Effective date: As of the date of issuance and shall be implemented within 90 days.

Amendment Nos.: Unit 1-169; Unit 2-170.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 9, 2003 (68 FR 68671). The supplemental letter dated January 7, 2004, provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 12, 2004.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendment: July 31, 2002 (superseded November 24, 1999, application) and its supplements dated August 15 and December 23, 2003.

Brief description of amendments: The amendments revise the technical specifications to relocate the pressure-temperature limits and low temperature overpressure protection system limit setpoints into a plant-specific pressure temperature limits report that will be administratively controlled by the technical specifications.

Date of issuance: May 13, 2004.

Effective date: May 13, 2004, and shall be implemented within 30 days from the date of issuance.

Amendment No.: Unit 1-170; Unit 2-171.

Facility Operating License Nos. DPR-80 and DPR-82: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 17, 2002 (67 FR 58648). The supplemental letters dated August 15 and December 23, 2003, provided additional clarifying information, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 13, 2004.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: February 2, 2004.

Brief description of amendments: The amendments revised the Technical Specification 3.1.8, “Scram Discharge Volume (SDV) Vent and Drain Valves,” for the condition of having one or more SDV vent or drain lines with one valve inoperable.

Date of issuance: May 25, 2004.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 240 and 183.

Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 16, 2004 (69 FR 12372). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 25, 2004.

No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendment: March 23, 2004, as supplemented April 30, 2004.

Brief description of amendment: The amendments allow both trains of control room air-conditioning system (CRACS) to be inoperable for up to 7 days provided control room temperatures are verified every 4 hours to be less than or equal to 90 degrees Fahrenheit. If this temperature limit cannot be maintained or both CRACS trains are inoperable for more than seven days, the requirements of Technical Specification Section 3.0.3 must be implemented.

Date of issuance: May 21, 2004.

Effective date: As of the date of issuance and shall be implemented within 30 days of issuance.

Amendment No.: 292 and 282.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments revised the technical specifications.

Date of initial notice in Federal Register: April 14, 2004 (69 FR 19880). The April 30, 2004, letter provided clarifying information that did not expand the scope of the original application or change the initial proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 21, 2004.

No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North Anna Power Station, Units 1 and 2, Louisa County, Virginia

Date of application for amendment: December 13, 2002, as supplemented by letters dated May 8, 2003, December 17, 2003, February 12, 2004, and March 9, 2004.

Brief description of amendment: These amendments revise the completion time of Required Action A.1 of Technical Specification 3.8.7, “Inverters—Operating,” from 24 hours to 7 days for an inoperable instrument bus inverter.

Date of issuance: May 12, 2004.

Effective date: As of the date of issuance and shall be implemented within 90 days from the date of issuance.

Amendment Nos.: 235 and 217.

Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: April 15, 2003 (68 FR 18289). The May 8, 2003, December 17, 2003, February 12, 2004, and March 9, 2004, supplementary letters contained clarifying information only and did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated May 12, 2004.

No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and Final Determination of No Significant Hazards Consideration and Opportunity for a Hearing (Exigent Public Announcement or Emergency Circumstances)

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the date the amendment was needed, there was not time for the Commission to publish, for public comment before issuance, its usual Notice of Consideration of Issuance of Amendment, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a Federal Register media to provide notice to the public in the area surrounding a licensee's facility of the licensee's application and of the Commission's proposed determination of no significant hazards consideration. The Commission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of communication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public comments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards consideration determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. If comments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public inspection at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendment. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, contact the PDR Reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of flaw or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner/requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Each contention shall be given a separate numeric or alpha designation within on of the following groups:

1. Technical—primarily concerns/issues relating to technical and/or health and safety matters discussed or referenced in the applications.

2. Environmental—primarily concerns/issues relating to matters discussed or referenced in the environmental analysis for the applications.

3. Miscellaneous—does not fall into one of the categories outlined above.

As specified in 10 CFR 2.309, if two or more petitioners/requestors seek to co-sponsor a contention, the petitioners/requestors shall jointly designate a representative who shall have the authority to act for the petitioners/requestors with respect to that contention. If a petitioner/requestor seeks to adopt the contention of another sponsoring petitioner/requestor, the petitioner/requestor who seeks to adopt the contention must either agree that the sponsoring petitioner/requestor shall act as the representative with respect to that contention, or jointly designate with the sponsoring petitioner/requestor a representative who shall have the authority to act for the petitioners/requestors with respect to that contention.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing. Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 415-1101, verification number is (301) 415-1966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer or the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: May 7, 2004.

Brief description of amendment: The amendment restores the licensed thermal power from 1524 megawatts thermal (MWt), as approved in Amendment No. 224, to the previous value of 1500 MWt.

Date of issuance: May 14, 2004.

Effective date: May 14, 2004.

Amendment No.: 227.

Renewed Facility Operating License No. DPR-40: The amendment revised the Operating License and the Technical Specifications. Public comments requested as to proposed no significant hazards consideration (NSHC): Yes. Omaha-World Herald. The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received.

The Commission's related evaluation of the amendment, finding of exigent circumstances, State consultation, and final NSHC determination are contained in a safety evaluation dated May 14, 2004.

Attorney for licensee: James R. Curtiss, Esq. Winston & Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Section Chief: Stephen Dembek.